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Optical diffusion is an essential process used to manage photons in a wide range of photoelectric systems. This work proposes an approach to fabricate novel optical diffusers by a plasma-processing technique, using helium plasma to form fiberform nanostructures on tungsten thin films. After an annealing procedure in the air for oxidation, the optical properties and the light-diffusing abilities of these nanostructured thin films were studied. The diffusion efficiency of the optical diffusers was analyzed using a transmitted scatter distribution function (TDF), and it was revealed that the diffusion efficiency of a device with an irradiation time of 30 minutes could reach 97%.

Hydrogen isotope retention and chemical state for the tiles exposed to plasma in the JET–ITER-like wall (ILW) during two campaigns in 2011–2012 (first campaign, ILW-1) and 2015–2016 (third campaign, ILW-3) were studied and compared by means of X-ray photoelectron spectroscopy and thermal desorption spectroscopy. In both campaigns the upper part of the inner divertor tiles was the deposition-dominated area, while erosion was observed on the outer divertor tiles. Therefore, higher deuterium retention was found on the inner divertor tiles. The major D desorption peak for the inner divertor tiles from ILW-3 was located at the temperature range of 470°C to 520°C, which was higher than measured after ILW-1: 370°C to 430°C. The XPS analyses showed the formation of a BeO layer on the ILW-3 inner divertor tiles, while after ILW-1 the layers also contained a significant amount of carbon. Deuterium retention was reduced toward the outer divertor tiles. The differences could be related to the difference in the power level in the two campaigns.

A fiberform nanostructure was synthesized by exposing high-density helium plasma to a 100 nm thick tungsten thin film in the linear plasma device NAGDIS-II. After helium plasma exposure, the cross-section of samples was observed by a scanning electron microscope, transmission electron microscope, and focused ion beam scanning electron microscope. It is shown that the thickness of the nanostructured layer increases significantly for only a short irradiation time. The optical absorptivity remains high, even though it is exposed to helium plasma for a short time. The usage of the thin film can shorten the processing time for nanostructure growth, which will be beneficial for commercial production.

Ceramic coatings have been investigated as tritium permeation barrier for several decades; however, there is little knowledge about irradiation effects on the coatings. In this study, yttrium oxide coatings were fabricated on reduced activation ferritic/martensitic steel substrates by radio frequency reactive magnetron sputtering, and their microstructures were investigated after iron-ion irradiation under various parameters with damage concentration of up to 20 displacement per atom. The formation of voids and an amorphous layer was confirmed in all the irradiated samples. In addition, it was found that irradiation-induced grain growth proceeded with damage concentration, and the increase of grain size was proportional to damage concentration. Furthermore, a layer with few voids formed between the crystallized and amorphous layers for the irradiated samples with higher incident energy or higher dose. Defect recovery and heat-driven grain growth and/or crystallization occurred for the sample irradiated at 500 °C, resulting in a larger average grain size than those irradiated at room temperature. Further consideration of irradiation effects to the coatings under fusion-relevant irradiation conditions is necessary.

To investigate the surface damage of a material by He ions, a dispersion-strengthened W-1wt%TiC alloy was irradiated by 5-keV He ions at 773 K, 973 K, and 1173 K up to an ion dose of 1.8 × 1021 He m−2, respectively. No He bubble formation was observed under transmission electron microscopy at any temperature for He doses less than 1.5 × 1020 He m−2. When this dose was exceeded, He bubbles grew and void swelling increased with increasing irradiation doses. Naturally, the growth of He bubbles and the void swelling became more pronounced with increasing irradiation temperatures. Compared to the published data on commercially available pure W, the formation of He bubbles was suppressed by TiC particles in the W–TiC alloy, especially at low doses.

In our previous work, the joint between oxide dispersion strengthened copper alloy (ODS-Cu), GlidCop® (Cu–0.3 wt%Al2O3) and tungsten (W) demonstrated superior fracture strength (∼200 MPa). This joint was fabricated by the direct brazing method between W and ODS-Cu using BNi-6 (Ni–11%P) filler material without any intermediate layer. This method was named as the improved or the advanced brazing technique. In the present study, deformation and fracture behavior of the joint after the three-point bending test was investigated. At first, it was found that the crack initiation points were dominantly in the W bulk, although it was not clear that the crack initiated from grain boundary or not. Secondly, the crack propagation proceeded mostly in the W bulk but tended to deflect towards the bonding layer. These results interpret the strength of the bonding interface is superior and the present bonding technique is applicable for severe environments such as a high heat flux divertor component on the fusion reactor. Based on the above physical and technological understanding, we successfully fabricated the large scale divertor mock-up which has twenty-eight plates of W with each size of 20 × 20 × 5 mm3.

A laboratory-scale inductively coupled plasma apparatus investigated the behavior of liquid Sn–Bi–Li–Er alloy (SBLE) under hydrogen (H2) plasma exposure. By exposing the liquid SBLE to H2 plasma, the bubbling of liquid SBLE and pulsive H2 pressure increase was observed. Moreover, plasma density, i.e. ion flux, increased both bubble size and H2 pressure spike. The simultaneous measurement of a high-speed camera and mass spectrometry concluded that the bubble was formed by H2 accumulation, and ∼38% of hydrogen ion flux contributed to the H2 accumulation. Lithium hydride may be considered the bubble sphere’s origin from X-ray photoelectron spectroscopy measurement of droplets produced from the ruptured bubble sphere. From the experimental results, the bubble formation mechanism was discussed.

In the Large Helical Device (LHD), the first deuterium plasma experiment was conducted in 2017. To investigate tritium migration in the LHD vacuum vessel, long-term material probes were installed on the first wall before the deuterium plasma experiment. After the experiment, the microstructure and amount of tritium remaining in each probe were analyzed. The results showed that a relatively large amount of tritium remained in the probes on the first wall, forming a thick deposition layer, rather than in the probes located in the erosion-dominant area. In the deposition layers on the probes, the dominant element is carbon, which can be generated on the divertor tiles made of graphite. The result of orbit calculation of the energetic tritons in the case of the standard magnetic configuration in the LHD showed that approximately 40% of the tritons generated by deuterium–deuterium fusion reactions were promptly lost mainly to the divertor. Thermalized tritons also flew to the divertor along with the background plasma. The divertor tiles, on which the tritons impinged, were eroded by the divertor plasma, and carbon atoms and tritiated hydrocarbon molecules were generated and deposited on the first wall. This can be the dominant mechanism of tritium retention in the first wall. Among the material probes located in the erosion-dominant area, the amount of tritium remaining in the probe on which the energetic tritons impinged was relatively large. The results of the tritium balance analysis show that the first wall is not the dominant reservoir of tritium in the LHD.

Tritium (T) distribution on the plasma-facing surfaces (PFSs) and inside castellation of Be limiter tiles from the JET tokamak with the ITER-like wall (ILW) was analyzed using imaging plate (IP) technique and β-ray induced X-ray spectrometry (BIXS). Regarding to PFSs, the outer poloidal limiter (OPL) showed significantly higher T concentrations than the inner wall guard limiter (IWGL) and upper dump plate (DP). The concentration of T on OPL was high at the central part. However, deuterium (D) and metallic impurities showed maximum concentration at the edges. This difference in distributions indicated different deposition and retention mechanisms between T and D. In contrast, deposition profiles of T concentrations on the castellated surfaces extended up to ∼ 5 mm into the gap, i.e. were similar to those of D and metallic impurities found by ion beam analysis.

Large-scale mm thick fiberform nanostructures are grown when materials are subjected to high flux helium ions with tungsten particles (ions and neutrals) precipitation. Observations of the large-scale fiberform nanostructures (LFNs) were conducted by scanning electron microscope (SEM) and transmission electron microscope (TEM) analyses. In particular, this study focuses on membrane nanostructures. Membranes that contain He bubbles were always observed when LFNs covered the surface. We discuss the growth process of the membrane structures based on detailed observations. Keywords: Nanostructures, Helium, Tungsten, Transmission electron microscope

Three sets of tungsten coated graphite divertor tiles (VPS-W tiles) were installed in the closed helical divertor of the Large Helical Device (LHD) instead of the graphite divertor tiles in the 2012FY plasma campaign for examining the plasma wall interaction (PWI) with divertor plasma. The first wall panels and divertor tiles of the LHD consist of stainless steels (SUS316L) and graphite, respectively. The carbon based mixed-material deposition layer including a very small amount of Fe element has been formed on not only the divertor tiles but also the first walls near the divertor tiles through the PWI processes. Such a mixed layer often causes undesirable influences for maintaining a long pulse discharge in LHD, e.g., changing the particle recycling properties and dust generation. After the single plasma campaign of the 2012FY, we confirmed drastic suppression of the mixed-material deposition layer on the first wall panels just under the VPS-W tiles. On the other hand, carbon based mixed-material deposition layer was formed on the VPS-W tile surface, where the amount of the hydrogen retention was estimated to be over ∼4 × 1021 H/m2. If we would want further suppression of the hydrogen retention on the VPS-W tiles, the plasma facing components should be replaced to a full metal wall to avoid formation of the carbon co-deposition layer. Keywords: Tungsten divertor, Dust, Hydrogen retention

Erbium oxide has been investigated as a tritium permeation barrier material in a D-T fusion reactor fuel system for more than a decade. Erbium oxide normally forms the cubic phase; however, the uncommon monoclinic phase was formed under ion irradiation or high pressure conditions. In this study, the monoclinic erbium oxide coatings were prepared to examine their microstructure and deuterium permeability with ion-irradiation effect. The monoclinic phase coatings with a preferred orientation showed 1–2 orders of magnitude higher permeability at 300–400 °C than the cubic phase coating reported previously due to the smaller grain with uncrystallized region. After the permeation measurement at 600 °C, the permeability drastically decreased due to the phase transformation to the cubic phase and a change in grain structure from columnar to granular. An Fe-ion irradiated coating with the damage concentration of 0.05 dpa showed the decrease in the permeability at 500 °C, indicating the irradiation damage to the grain structure may accelerate the grain growth at lower temperature, while the phase transformation would occur at 600 °C because the diffusivity did not change much at 500 °C and drastically decreased at 600 °C.

To understand the fuel retention mechanism, correlation of surface chemical states and hydrogen isotope retention behavior determined by XPS (X-ray photoelectron spectroscopy) and TDS (Thermal desorption spectroscopy), respectively, for JET ITER-Like Wall samples from operational period 2011-2012 were investigated. It was found that the deposition layer was formed on the upper part of the inner vertical divertor area. At the inner plasma strike point region, the original surface materials, W or Mo, were found, indicating to an erosion-dominated region, but deposition of impurities was also found. Higher heat load would induce the formation of metal carbide. At the outer horizontal divertor tile, mixed material layer was formed with iron as an impurity. TDS showed the H and D desorption behavior and the major D desorption temperature for the upper part of the inner vertical tile was located at 370 °C and 530 °C. At the strike point region, the D desorption temperature was clearly shifted toward higher release temperatures, indicating the stabilization of D trapping by higher heat load.

Dispersion-strengthened W alloys, W-1%TiC and W-2%TiC, were fabricated by spark plasma sintering (SPS). The thermal stabilities of the secondary-phase TiC particles and microstructures in the W alloys during thermal treatments from 100 to 1000 °C were investigated by positron annihilation techniques. The TiC particles were stable in both W–TiC alloys during the annealing process. Both alloys were irradiated by 5-keV He ions at 900 °C–1.8 × 1021 He/m2. Compared with commercially available pure W, the formation of He bubbles was suppressed by the TiC particles in the W–TiC alloys, and the extent of suppression increased as the amount of added TiC was increased. Keywords: W-TiC, Thermal stability, He irradiation, He bubble

Growth of helium bubbles near the surface of metals leads to various morphology changes when sufficient helium ions are continuously implanted on the surface. In this study, low energy (< 100 eV) helium plasma irradiations to various noble metals were conducted, and consequent surface morphology changes were analyzed. It was found that fiberform nanostructures were formed by the helium plasma irradiation on rhodium and ruthenium thin films, which were deposited by magnetron sputtering, and on platinum and iridium wires. Growth of fiberform structures were not identified on gold, silver, and palladium samples exposed to helium plasmas even though the irradiations were conducted at various surface temperatures, though pinholes were observed on palladium surface. We discussed the relation between the morphology changes and material properties., ファイル公開：2020-04-25

The purpose of this study is to elucidate helium (He) retention behavior in tungsten (W) under simultaneous He and hydrogen (H) irradiation. Polycrystalline-W was irradiated by He+ and H2+ simultaneously with the energy of 1.0 keV and 3.0 keV. He+ fluences were (0.5, 1.0, 10) × 1021 He+ m−2 and H2+ fluence was 1.0 × 1022 H+ m−2,respectively. After irradiation, He desorption behavior was investigated by high temperature thermal desorption spectroscopy (HT-TDS) in the temperature range of R.T.-1773 K. Micro-structure changes of W after irradiation were observed by TEM. It was found that simultaneous irradiation with different H2+ energy significantly changed He retention behavior. 1.0 keV H2+ suppressed the He bubble growth and no bubbles can be observed at room temperature. On the other hand, 3.0 keV H2+ facilitated the formation of He bubbles and increased the He retention due to the additional damage introduction by energetic H2+.

Fiberform nanostructured tungsten (FN-W) samples were synthesized by helium plasma irradiation with different irradiation times. It is shown that the optical reflectivity decreases significantly with the increase of the irradiation time, even in a long-wavelength infrared range up to several dozen µm. These experimental results are of importance for the usage of retroflectors and will offer a promising direction for other practical applications.

The reliable bonding between tungsten (W) and reduced activation ferritic/martensitic (RAFM) steel (JLF-1) was obtained by brazing with BNi-6 (Ni-11%P) filler material. In this work, a pure copper (Cu) interlayer was selected for the absorber of the residual stress, in which BNi-6 filler materials are inserted between W/Cu and Cu/JLF-1 interface. The stacked structure of the W/BNi-6/Cu/BNi-6/JLF-1 was subjected to the heat treatment procedure at 960℃. After the heat treatment procedure, fine joint structure without any crack and large size pore can be confirmed. In addition, several kinds of detailed analysis such as hardness testing and element distribution in the W/BNi-6/Cu/BNi-6/JLF-1 bonding structure, were performed. It was found that the Ni and P elements were preferentially located near each joint interface, and those elements caused a hardening effect of the bonding layer through some segregation mechanism. Also, the joint sample is subjected to the heat loading test by the electron beam device (ACT2) for evaluating the heat conducting characteristics of the joint. According to the temperature measurement during the heat loading, those hardening layers seemed to not cause the negative effects for degrading the joint properties.

In this study, global particle balance of helium (He) long pulse discharge with graphite divertor in the Large Helical Device (LHD) and He retention in graphite from lab-scale experiments were compared in order to determine the role of the graphite divertor in He retention on the LHD wall. Global He particle balance analysis was conducted in long-pulse discharges in LHD with only turbo molecular pump. The analysis showed that static He retention was ∼2.9 × 1022 He and the ratio of retained He (= retention over fluence) was ∼0.45%. In the lab-scale study, He retention was measured by Thermal Desorption Spectroscopy (TDS) after ion irradiation or plasma exposure. The ratio of retained He were 0.1 ∼ 0.5% in all conditions, which was well consistent with LHD results. Therefore, it may be concluded that graphite divertor tiles have an important role in He absorption in the LHD wall at the initial phase of discharges.

A dispersion-strengthened W alloy, W-La 2 O 3 , was developed by spark plasma sintering (SPS). The D 2 retention and microstructural evolution under 5-keV He irradiation at high temperatures were investigated. Although the density of the W-La 2 O 3 alloy was lower than that of commercially available W, the D 2 retention did not change significantly. In addition, the D 2 thermal desorption peak, which did not appear in commercial W, appeared at ∼700 K due to the dissociation from D to vacancy clusters. In W-La 2 O 3 , the formation of He bubbles was observed at 773 K, 973 K and 1173 K. As the irradiation temperature was increased, void swelling also increased. For the same temperature, the formation of He bubbles was easier for W-La 2 O 3 than for commercial W. This is considered to be the cause of low density of W-La 2 O 3 . W-La 2 O 3 had a higher density of vacancies before He irradiation. The suppression of vacancy aggregation induced by the dispersion of La 2 O 3 as the second phase appeared only at a high irradiation doses.

Influence of mixed material layer formation on hydrogen isotope retention in W exposed to 2014 Large Helical Device (LHD) experiment campaign was evaluated by Thermal Desorption Spectroscopy (TDS), X-ray Photoelectron Spectroscopy (XPS), Scanning Electron Microscope (SEM) and Transmission Electron Microscope (TEM). It was found that a lot of hydrogen isotopes were trapped by the carbon-dominated mixed-material layer deposited on the plasma facing materials. Most of He was also trapped in the carbon-dominated mixed-material layer and the corresponding desorption temperature was limited to be about 600 K, 900 K and 1200 K, respectively. However, the hydrogen retention behavior for erosion dominated area was clearly different from those for deposition dominated area and typical Plasma Wall Interaction (PWI) area, where He bubbles were introduced near the sample surface, leading to the introduction of various types of trapping sites in W.

Surface morphology and hydrogen isotope retention of W specimen melted with vertical displacement event-like heat load and subsequent deuterium (D) plasma exposure were studied. Applied heat loads using electron beam without raster scanning were about 190 and 230 MW/m2 in heat flux and 0.08, 0.12 and 0.16s in duration. After the heat load application, specimens showed apparent melting spots with grain growth or dense micrometer scale convex structure. Cracks were observed only in the part with the convex structure. D retention in the melted part of specimens was not significantly larger than in the reference specimen despite large changes of surface characteristics. Keywords: Tungsten, Retention, Melting, Divertor, High heat flux, NRA

In order to confirm the formation of a tungsten fiberform nanostructure (W-fuzz) by helium plasma exposure in the large-sized plasma confinement device, the bulk tungsten with the size of 80 × 20 × 1.5 mm3 was inserted into the divertor leg position in the Large Helical Device (LHD). Then, it was exposed to the divertor plasma during the ultra-long pulse helium discharges with 10,190s in total. The width of the divertor plasma, incident ion energy and total fluence were expected to be ∼2cm, 100–200eV and ∼5 ×1025 He/m2 (strike position), respectively. The surface temperature of the tungsten specimen was monitored by IR camera. The typical surface temperature of the divertor strike point was estimated to be around 1900∼2300K. After the exposure, an initial growth phase of tungsten fiberform nanostructure (W-fuzz) was able to be identified on the tungsten surface. The finest initial growth phase of the W-fuzz structure was able to be identified on the central region of the divertor strike point, where retention amount of helium was estimated to be ∼8 ×1021 He/m2. This study is the first simultaneous evaluation of the W-fuzz growth and quantification of the helium retention in the large-sized plasma confinement device. Keywords: W-fuzz, TEM observation, Ion beam analysis, LHD

All the hydrogen isotope (H, D, T) simultaneous TDS (Thermal desorption spectroscopy) measurement system (HI-TDS system) was newly designed to evaluate all hydrogen isotope desorption behavior in materials. The present HI-TDS system was operated under Ar purge gas and the H and D desorptions were observed by a quadruple mass spectrometer equipped with an enclosed ion source, although T desorption was evaluated by an ionization chamber or proportional counters. Most of the same TDS spectra for D and T were derived by optimizing the heating rate of 0.5 K s−1 with Ar flow rate of 13.3 sccm.Using this HI-TDS system, D and T desorption behaviors for implanted or DT gas exposed tungsten samples installed in LHD (Large Helical Device) at NIFS (National Institute for Fusion Science) was evaluated. It was found that major hydrogen desorption stages consisted of two temperature regions, namely 700 K and 900 K, which was consistent with the previous hydrogen plasma campaign and most of hydrogen would be trap...

A new temperature controlled material probe was designed for the exposure of tungsten samples to helium plasma in the LHD. Samples were exposed to estimated fluences of ∼1023 m−2 and temperatures ranging from 65 to 600 °C. Transmission Electron Microscopy analysis allowed the study of the impact of He irradiation under high temperatures on tungsten micro structure for the first time in real-plasma exposure conditions. Both dislocation loops and bubbles appeared from low to medium temperatures and saw an impressive increase of size (factor 4 to 6) most probably by coalescence as the temperature reaches 600 °C, with 500 °C appearing as a threshold for bubble growth. Annealing of the samples up to 800 C highlighted the stability of the dislocation damages formed by helium irradiation at high surface temperature, as bubbles and dislocation loops seem to conserve their characteristics. Additional studies on cross-sections showed that bubbles were formed much deeper (70–100 nm) than the heavily damaged surface layer (10–20 nm), raising concern about the impact on the material mechanical properties conservation and potential additional trapping of hydrogen isotopes.

Accelerator-driven neutron sources have been developed worldwide for boron neutron capture therapy (BNCT) instead of nuclear reactors. Nagoya University is currently developing a compact DC accelerator driven BNCT system using a sealed lithium (Li) target. The target receives a high heat load by a proton beam, and the heat then requires efficient removal. This study developed a high-efficiency heat removal technique that uses V-shaped staggered ribs on the heat removal side of the flow channels. We performed experiments using an electron beam to confirm the technique performance. The results showed double the heat removal performance compared to the smooth surface channel. The value expected to sufficiently remove 6.6 MW/m2 of heat load provided by the proton beam at the Nagoya University neutron source for BNCT.

To precisely evaluate a butt joint technology for the JT-60SA CS coils, microstructure observations on the butt joint composed of Nb 3 Sn CIC conductors were conducted using a FE-SEM. As a sample for the observations, the butt joint sample utilized in the joint resistance measurement was used. During the sample fabrication, the butt joint sample was heated up to about 920 K from room temperature for diffusion bonding after heat treatment for Nb 3 Sn production. Then, the sample was subjected to the cycles of electromagnetic force in the joint measurement. The observation results indicated that Nb 3 Sn strands and a copper sheet were butted properly at the interface of the butt joint. In addition, there were hairline cracks in the Nb 3 Sn layers of the strands near the interface. To investigate a cause of the crack initiation, the stresses generated in the butt joint under same conditions were analyzed using a simple model. As a result, the cracks would occur with an axial compressive stress generated by the butt joint fabrication.

Optical reflectivity measurement is proposed as a convenient diagnostics method of surface modifications in plasma facing materials (PFMs) and its applicability to a plasma confinement device is evaluated. The optical reflectivity for mirror polished stainless steel (SUS316L) exposed to the Large Helical Device (LHD) plasma was examined with a spectroscopic ellipsometer. In-situ measurement of the reflectivity change under the glow discharge was also performed using a super continuum white laser. The detectable change of the reflectivity was observed and the level of the degradation depended on the location, exposure period and employing wavelength. Microstructure observation of the mirror surface revealed that the behavior of the reflectivity change correlated closely with the ion-induced damages and the depositions thickness in the vicinity of the sample surface. Thus, the optical reflectivity measurement is considered to be a possible method for convenient in-situ diagnostics for PFMs.

Pure tungsten (W) samples were respectively installed on the two positions of typical plasma wall interaction area (PI) and erosion dominated position (ER) on the first wall in the Large Helical Device (LHD) at National Institute for Fusion Science (NIFS), Japan. After the experiment campaign in 2014, these samples were picked up and thermal desorption spectroscopy (TDS) were applied to evaluate their desorption behavior of He which was implanted by He discharge. It was found that a carbon-dominated mixed-material layer with impurities, such as Fe, Cr, Mo, O and N, etc., was deposited layer by layer during the plasma exposure on the PI sample. On the other hand, W-C mixed layer was formed on the ER sample. The results showed that a large amount of He was trapped in the samples on both PI and ER, and the total He retention for ER is about twice as large as that of PI. He was trapped in various types of trapping sites in the ER sample and their desorption peaks were located at temperatures of about 425 K, 755 K, 1130 K and 1630 K. For PI sample, most of He was trapped in the carbon-dominated mixed-material layer and the corresponding desorption temperature was limited to be about 600 K, 900 K and 1200 K. The additional 3.0 keV helium ion (He+) implantation was performed for several samples to investigate the He retention characteristics in these samples and it was found that no additional desorption stage was found. These results suggested that the He discharge history and its deposition on the W plasma facing wall would affect He desorption behavior of W.

During the 19th experimental campaign in the Large Helical Device (LHD) in 2017, deuterium (D) plasma experiments were performed for the first time in the LHD. Tritons were generated by the D-D reaction in the core plasma during the plasma discharges. To remove the tritium in plasma facing surfaces for safety in maintenance, hydrogen (H) plasma experiments were performed for one month after the four months of D plasma experiments. In this study, the removal effect of the tritium by H plasma discharges was investigated. Six stainless steel (SUS316L) samples were introduced into a position of the first-wall of the LHD in the last month of the D experiments, which were exposed to the D plasma for one month and then, three samples were continued to exposure to the following H plasma for another one month. The experimental results showed that one-half of the tritium in the surface region of the SUS sample was removed, which was measured by the TIPT (Tritium imaging plate technique). Further, hydrothermal treatments were performed and the results showed that the tritium retention in the bulk of the SUS materials was also reduced. Meanwhile, the carbon content in the near-surface region was also reduced after H plasma experiments, which reduces the possibility of the formation of the C H chemical bond.

A novel fabrication method for a divertor heat removal component with tungsten (W) armour and copper alloy heat sink was newly developed and named as the “Advanced Multi-Step Brazing (AMSB). The basic principle of multi-step brazing is to apply the “advanced brazing technique” repeatedly during the manufacturing process of the single divertor heat removal component. The advanced brazing technique was originally developed by our previous work for joining between oxide dispersion strengthened copper alloy (ODS-Cu) and W with BNi-6 (Ni–11%P) filler material. The applied ODS-Cu was GlidCop® (Cu-0.3wt%Al2O3). One of the possible examples of the AMSB fabrication process can be considered as follows. First, an appropriate cooling flow path channel is processed into a GlidCop® heat sink. Then, the flow path channel is sealed in a leak tightness condition with a lid made by GlidCop®. The leak tightness joint between GlidCop® (GlidCop®/GlidCop®) can be realized by application of the “advanced brazing technique.” Second, in order to facilitate the weldability to connect the other cooling pipe system, the sleeves made by stainless steel (SUS) are jointed on the interface edge of the flow path channel of the GlidCop® heat sink with a leak tightness condition by the “advanced brazing technique.” Finally, W armour is jointed on the GlidCop® heat sink also by the “advanced brazing technique.” In this study, the mechanical strength of the SUS/GlidCop® joint was confirmed by the three-point bending test. Then, AMSB divertor mock-up with leak tightness condition was successfully produced.

In a fusion reactor, strict control of tritium is required from the viewpoints of sufficient fuel balance and radiological hazard. Fabrication of tritium-impermeable ceramic coatings is a promising solution, and ceramic coatings have been investigated for several decades. In recent years, several irradiation tests have been conducted for the ceramic coatings to understand irradiation effects on their characteristics and functions. However, there are few experimental reports on multi-irradiation tests of neutrons and alpha particles produced by nuclear reactions and radioactive decay of tritium. In this study, yttrium oxide coatings were fabricated by reactive and non-reactive radio frequency magnetron sputtering, and their microstructures and deuterium permeation behaviors were investigated after heavy ion irradiation and helium implantation tests simulating irradiation damage by neutrons and generated alpha particles. All irradiated samples showed the formation of an amorphous layer and a nanocrystalline layer or region. In particular, an iron-helium-simultaneous-irradiated sample contained voids including nanometer-size spheres, indicating the formation of helium bubbles. The irradiated samples showed grain growth and/or crystallization at a higher temperature than the unirradiated samples during deuterium permeation tests. That suggests that the voids and the helium bubbles prevented structural change at low temperatures. Moreover, the helium bubbles tended to be more stable than the voids. On the other hand, nickel-single and nickel-helium-sequential-irradiated coatings deposited by non-reactive sputtering showed few voids and helium bubbles, resulting in a different behavior of deuterium permeation.

Forming nanocone structures on a silicon (Si) surface by low (

In this study, the effect of ELM-like thermal shocks on He induced surface morphology were investigated. A W sample was exposed to pure He plasma. He ion incident energy, flux and fluence were 80 eV, 1 × 1022 /m2/s and 3 × 1025 /m2, respectively. Irradiation temperature were approximately 470 and 1100 K. Then, thermal shocks were applied on the sample using focused electron beam. The peak heat flux, pulse duration and base temperature were 500 MW/m2, 500 μs and R.T., respectively. After the thermal shocks, He induced morphologies such as holes and periodic undulations were flatten completely at the region exposed to the highest heat flux. At the peripheral regions of the electron beam spot, the hole density increase or partial flattening of morphologies were observed. These results suggested that in order to anticipate surface morphology with He irradiation and ELMs, the peak temperature should play a more important role than base temperature. Keywords: Tungsten, Divertor, He induced surface morphology, Thermal shocks

In the stellarator Wendelstein 7-X with its twisted 3D magnetic field geometry, studies of material migration with respect to first wall components become very important in view of the envisioned long-pulse operation. A variety of erosion/deposition samples were installed on the plasma-facing components exposed at three different nominal heat load levels between 0.1 and 10 MW/m2. After the first successful operation phase in divertor configuration, all the probes at higher and lower load levels were removed, whereas at the intermediate load levels, 352 out of 30 000 screws have been exchanged at selected locations along the toroidal and poloidal directions. The exchanged probes have been analyzed by various measurement techniques. At the higher load levels where the probes were installed within the divertor, heavy erosion has been observed presumably at the strike line positions. Both, erosion and deposition phenomena have been found on the screw heads. The optical reflection measurement profile of the whole plasma vessel show the deposition patterns at similar locations in all the five modules. At the low load level, the Si-wafer probes are under investigation.

Micro/nanoscopic observations on the surface of the divertor tiles used in the first campaign (2011-2012) of the JET tokamak with ITER-like Wall (JET ILW) have been carried out by means of several material analysis techniques. Previous results from the inner divertor were reported for a single poloidal section of the tile numbers 1, 3 and 4, i.e., upper, vertical and horizontal targets, respectively. The formation of the thick stratified mixed-material deposition layer on tiles 1 and 4, and erosion on tile 3 were identified. This study is mostly focused on the outer divertor: tiles 6, 7 and 8. In contrast to the inner tile, remarkable surface modifications have not been observed on the vertical target (tiles 7 and 8) where sputtering erosion and impurity deposition would have been almost balanced. Only a specific part of tile 6 (horizontal target) located near the exhaust channel was covered with a stratified ("geological-like") mixed-material deposition layer which mainly included Be and Ni with the thickness of similar to 2 mu m. Special feature of this mixed layer was that a certain amount of nitrogen (N) was clearly detected in the layer. Since the concentration of N varied with the depth position, it could be depended on the amount of that gas puffed for plasma edge cooling during the JET experimental campaign. In addition to the outer divertor tiles, a very interesting feature of the local erosion and deposition effects is reported in this paper.

Tritium permeation through structural materials is a critical issue in fusion reactors from the viewpoints of sufficient fuel balance and radiological hazard. Ceramic coatings have been investigated as tritium permeation barrier for several decades; however, irradiation effects of the coatings on permeation are not elucidated. In this work, yttrium oxide coatings were fabricated on reduced activation ferritic/martensitic steels by radio frequency magnetron sputtering, and their microstructures and deuterium permeation behaviors were investigated before and after iron-ion irradiation at different temperatures. An as-deposited coating had a columnar structure and transformed into a granular one after annealing. An amorphous layer formed near the coating-substrate interface of irradiated coatings, and its thickness became thinner with increasing irradiation temperature. Voids of approximately 20 nm in diameter also formed in the irradiated coatings. Deuterium permeation flux of the sample irradiated to 1 dpa at room temperature was the lowest among the unirradiated and irradiated samples, and a permeation reduction factor indicated up to 390. The amorphous layer disappeared after deuterium permeation measurements due to damage recovery, while the voids remained and aggregated. The irradiation damage would accelerate nucleation of the crystal, resulting in a decrease of the permeation flux.